The workshop is organized by the Nuclear Physics Institute of Academy of Sciences of the Czech Republic, public research institution, together with the Institute of Radiopharmaceutical Cancer Research of Helmholtz-Center Dresden-Rossendorf and in cooperation with the International Atomic Energy Agency (IAEA) and the support of many private sponsors. It is rather symbolic that Czech and German research institutions joined now freely their powers in order to organize this event.

Introduction 68Ge (t½ = 270.8 d, 100% EC) is an important radionuclide for two reasons: 1) once in equilib-rium with its daughter nuclide 68Ga (t½ = 68 min, 89 % β+, 3 % 1077 keV γ), it can be used as a positron source for attenuation correction and calibration of PET/MRI scanners; and 2) it can be employed as a generator of 68Ga for radiophar-maceutical preparation. Most isotope production facilities produce it using natural gallium (60.1% 69Ga, 39.9% 71Ga, melting point: 39 °C) as target material for proton bombardment at energies > 11.5 MeV, the threshold energy for 69Ga(p,2n)68Ge [1]. A maximum cross section of ~330 mb for natGa(p,x)68Ge occurs at ~20 MeV [1], hence proton energies in this neighborhood are mandatory for large scale production. Galli-um targetry is challenging due to its low melting point and corrosivity, hence compounds such as Ga2O3 (melting point: 1900 °C) or GaxNiy alloys (melting points > 800 °C) [2], have been used as target compounds [3,4,5]. The separation chem-istry technique employed by large-scale produc-tion facilities is liquid-liquid extraction using CCl4 [6,7]. In this work, two simple methods for GaxNiy alloy preparation are presented as well as a simple germanium separation procedure using a commercially available extraction resin. Material and Methods GaxNiy alloys were prepared by two methods (A,B). A) electrodeposition over 1.3 cm2 of a gold disk substrate. Ga2O3 and NiSO4.6H2O were dis-solved in a mixture of (27%) H2SO4 and NH4OH at pH 1.5 in a 3:2 mass ratio so that the Ga:Ni molar ratio was 4:1. The solution was then transferred to a 15 mL plating cell, in which a current of 29 mA/cm2 was applied with a platinum anode at 1 cm from the gold surface. B) Ga pellets were fused together with Ni powder at different Ga:Ni molar ratios using an induction furnace (EIA Power Cube 45/900). The resulting alloy pellets were then rolled to foils using a jeweler’s mill pressed between Nb foils to avoid contamination. Target irradiations were performed on a GE PETtrace at 16 MeV protons. The electroplated alloys were mounted on a custom-made solid target irradiation system with direct water-jet cooling applied to the backside of the gold disk. The alloy foils were placed on top of in a 1.2 cm diameter, 406 μm deep pocket made of Nb and sealed against a 51 μm Nb foil using a teflon O-ring. The alloys were in direct contact with the Nb foil to allow thermal conduction. At the rear of the Nb pocket is a water-cooling stream to transfer heat convectively during irradiation. Ge separation was achieved based on the difference in distribution coefficients between Ge, Ga, Zn, Cu, Ni and Co at different HNO3 molarities in DGA resin (Triskem International). Initial tests on the resin were performed after two pilot irradiations on natural gallium (a,b). a) 16 MeV protons were directed downward on an external beam-line (−30 °) onto 640 mg of molten elemental natGa pooled on a water-cooled niobium support. b) 330 mg natGa pellet was melted in the same Nb pocket well used with the alloys and was also sealed against a 51 μm Nb foil. The irradiated gallium was left to decay for 2 weeks and then was dissolved in 6 mL of concentrated HNO3. The solution was then passed through 200 mg of DGA resin packed in a 5 mm diameter column at a flow rate of 1.1 mL/min. A separation profile for Ge, Ga and Zn was obtained by collecting 0.2–1.0 mL fractions, which were analyzed by gamma ray spectroscopy on a HPGe detector. Two thick NiGa4 foils have been irradiated, one for 69Ge production and for radiocobalt, from 58Ni(p,α), separation quantification; and the other one for 68Ge production with the idea of preparing a mini-generator (< 13 MBq) of 68Ga for local use in phantom imaging work and animal studies. Results and Conclusion A) Each electroplating batch consisted of 66.5 ± 2.9 mg of Ga2O3 mixed with 44.9 ± 3.6 mg of NiSO4.6H2O (n = 9) in the 15 mL plating cell. Higher concentrations resulted in inefficient electroplating yields due to precipitation. 66 ± 6 % of the total Ga+Ni mass in solution, that is 39.5 ± 3.3 mg of Ga-Ni was deposited after 3 d. Three plating batches over one disk resulted in a maximum target thickness of 86.7 mg/cm2. A fourth batch did not add any significant amount of alloy and salt precipitation became a problem. The electroplated surface looked homogeneous at 10× magnification on a microscope and the targets were able to withstand up to 30 μA without presenting any dark spots. B) Alloys with Ga:Ni molar ratios of 1.0, 2.0, 2.9, 3.7 and 5.2 were fused by induction heating. TABLE 1 summarizes the results from manipulating these foils. These alloys were analyzed by X-ray fluorescence using a 109Cd excitation source quantifying the x-rays peaks: 9.26 keV for Ga and 7.48 keV for Ni. A linear relationship between the ratio of count rates of these two peaks to the alloy Ga:Ni molar ratio was found and employed for the characterization of the electroplated Ga-Ni layers. Results from the irradiations over natGa on Nb supports are presented in TABLE 2. TABLE 3 presents the results from irradiating two thick NiGa4 foils made by induction heating. Figure 1 contains the separation profile with DGA. 91% of the 68Ge is eluted in 2 mL of de-ionized water. We developed two simple methods for NiGa4 alloy manufacture. With a melting point > 800 °C and 80% presence of natGa, it is a more convenient target for 68Ge production compared to Ga encapsulated in Nb. The separation method based on the extraction resin DGA yields similar results as the liquid-liquid extraction method mentioned in [6,7], but we believe this is a more convenient method since it only requires a single trap-and-release step and not many extraction steps.

Introduction 52Mn (t½ =5.59 d, β+ = 29.6%, Eβmax = 0.58 MeV) has great potential as a long lived PET isotope for use in cell tracking studies, observation of immunologic response to disease states, or as an alternative to manganese-based MRI contrast agents. Its favorable max positron energy leads to superb imaging resolution, comparable to that of 18F.[1] Manganese is naturally taken up by cells via a multitude of pathways including the divalent metal transporter (DMT1), ZIP8, transferrin receptors (TfR), store-operated Ca2+ channels (SOC-Ca2+), and ionotropic glutamate receptor Ca2+ channels (GluR).[2] These natural transport mechanisms make 52Mn an attractive isotope for applications necessitating non-perturbative cell uptake. In particular, cell tracking is critical to the development and translation of stem cell therapies in regenerative medicine. Alternative-ly, 52Mn could be used in immunotherapy techniques such as adoptive cellular therapy (ACT) to evaluate the ability of external immune cells to reach their intended target. Material and Methods 52Mn was produced by natCr(p,x)52Mn using 16 MeV protons. The average thick target production yield was 0.23 mCi/µA-h with less than 0.25% co-production of 54Mn. Small amounts of 51Cr were observed in the target, but were absent from the radiochemically separated product. Target construction consisted of a water jet cooled chromium disc (3/4” diameter, 0.4” thick). Targets were purchased from Kamis Inc, and are 99.95% pure. Targets withstood beam currents of 30 µA with no visible aberration. Chromium targets were etched by concentrated HCl following bombardment. Mn2+ ions were extracted from 9M HCl to 0.8M trioctylamine in cyclohexane leaving the bulk chromium in the aqueous phase. After isolating the organic phase, 0.001M NH4OH was used to back-extract the Mn2+ ions to aqueous phase. This purification cycle was conducted a total of three times for each 52Mn production. Results and Conclusion For a starting bulk chromium mass of 456 ± 1 mg, a post-separation chromium mass of 5.35 ± 0.04 ng was measured by microwave plasma atomic emission spectrometry (MP-AES). This mass reduction corresponds to an average separation factor of 440 for a single purification cycle. Each purification cycle had a 52Mn recovery efficiency of 73 ± 7 % (n = 6), resulting in an overall separation efficiency of approximately 35 %. These efficiencies and separation factors agree reasonably well with the work conducted by Lahiri et. al.[3] Prior to use, the product was passed through a C-18 Sep-Pak to remove any residual organic phase. After four target irradiations and etchings, some pitting became noticeable on the target face. These have not yet compromised the o-ring seal with the target deplater, but it is possible that target replacement after every 6–9 52Mn productions will be necessary moving forward. Following the successful separation of 52Mn from chromium, in vitro experiments were conducted to demonstrate the uptake of 52Mn by human stem cells and mouse tumor cells. A linear uptake response was observed as a function of the amount of activity exposed to the cells for both cell models. These experiments have shown great promise for 52Mn as a long-lived PET isotope in cell tracking studies. Details will be presented.

The workshop is organized by the Nuclear Physics Institute of Academy of Sciences of the Czech Republic, public research institution, together with the Institute of Radiopharmaceutical Cancer Research of Helmholtz-Center Dresden-Rossendorf and in cooperation with the International Atomic Energy Agency (IAEA) and the support of many private sponsors. It is rather symbolic that Czech and German research institutions joined now freely their powers in order to organize this event.

Introduction 68Ge (t½ = 270.8 d, 100% EC) is an important radionuclide for two reasons: 1) once in equilib-rium with its daughter nuclide 68Ga (t½ = 68 min, 89 % β+, 3 % 1077 keV γ), it can be used as a positron source for attenuation correction and calibration of PET/MRI scanners; and 2) it can be employed as a generator of 68Ga for radiophar-maceutical preparation. Most isotope production facilities produce it using natural gallium (60.1% 69Ga, 39.9% 71Ga, melting point: 39 °C) as target material for proton bombardment at energies > 11.5 MeV, the threshold energy for 69Ga(p,2n)68Ge [1]. A maximum cross section of ~330 mb for natGa(p,x)68Ge occurs at ~20 MeV [1], hence proton energies in this neighborhood are mandatory for large scale production. Galli-um targetry is challenging due to its low melting point and corrosivity, hence compounds such as Ga2O3 (melting point: 1900 °C) or GaxNiy alloys (melting points > 800 °C) [2], have been used as target compounds [3,4,5]. The separation chem-istry technique employed by large-scale produc-tion facilities is liquid-liquid extraction using CCl4 [6,7]. In this work, two simple methods for GaxNiy alloy preparation are presented as well as a simple germanium separation procedure using a commercially available extraction resin. Material and Methods GaxNiy alloys were prepared by two methods (A,B). A) electrodeposition over 1.3 cm2 of a gold disk substrate. Ga2O3 and NiSO4.6H2O were dis-solved in a mixture of (27%) H2SO4 and NH4OH at pH 1.5 in a 3:2 mass ratio so that the Ga:Ni molar ratio was 4:1. The solution was then transferred to a 15 mL plating cell, in which a current of 29 mA/cm2 was applied with a platinum anode at 1 cm from the gold surface. B) Ga pellets were fused together with Ni powder at different Ga:Ni molar ratios using an induction furnace (EIA Power Cube 45/900). The resulting alloy pellets were then rolled to foils using a jeweler’s mill pressed between Nb foils to avoid contamination. Target irradiations were performed on a GE PETtrace at 16 MeV protons. The electroplated alloys were mounted on a custom-made solid target irradiation system with direct water-jet cooling applied to the backside of the gold disk. The alloy foils were placed on top of in a 1.2 cm diameter, 406 μm deep pocket made of Nb and sealed against a 51 μm Nb foil using a teflon O-ring. The alloys were in direct contact with the Nb foil to allow thermal conduction. At the rear of the Nb pocket is a water-cooling stream to transfer heat convectively during irradiation. Ge separation was achieved based on the difference in distribution coefficients between Ge, Ga, Zn, Cu, Ni and Co at different HNO3 molarities in DGA resin (Triskem International). Initial tests on the resin were performed after two pilot irradiations on natural gallium (a,b). a) 16 MeV protons were directed downward on an external beam-line (−30 °) onto 640 mg of molten elemental natGa pooled on a water-cooled niobium support. b) 330 mg natGa pellet was melted in the same Nb pocket well used with the alloys and was also sealed against a 51 μm Nb foil. The irradiated gallium was left to decay for 2 weeks and then was dissolved in 6 mL of concentrated HNO3. The solution was then passed through 200 mg of DGA resin packed in a 5 mm diameter column at a flow rate of 1.1 mL/min. A separation profile for Ge, Ga and Zn was obtained by collecting 0.2–1.0 mL fractions, which were analyzed by gamma ray spectroscopy on a HPGe detector. Two thick NiGa4 foils have been irradiated, one for 69Ge production and for radiocobalt, from 58Ni(p,α), separation quantification; and the other one for 68Ge production with the idea of preparing a mini-generator (< 13 MBq) of 68Ga for local use in phantom imaging work and animal studies. Results and Conclusion A) Each electroplating batch consisted of 66.5 ± 2.9 mg of Ga2O3 mixed with 44.9 ± 3.6 mg of NiSO4.6H2O (n = 9) in the 15 mL plating cell. Higher concentrations resulted in inefficient electroplating yields due to precipitation. 66 ± 6 % of the total Ga+Ni mass in solution, that is 39.5 ± 3.3 mg of Ga-Ni was deposited after 3 d. Three plating batches over one disk resulted in a maximum target thickness of 86.7 mg/cm2. A fourth batch did not add any significant amount of alloy and salt precipitation became a problem. The electroplated surface looked homogeneous at 10× magnification on a microscope and the targets were able to withstand up to 30 μA without presenting any dark spots. B) Alloys with Ga:Ni molar ratios of 1.0, 2.0, 2.9, 3.7 and 5.2 were fused by induction heating. TABLE 1 summarizes the results from manipulating these foils. These alloys were analyzed by X-ray fluorescence using a 109Cd excitation source quantifying the x-rays peaks: 9.26 keV for Ga and 7.48 keV for Ni. A linear relationship between the ratio of count rates of these two peaks to the alloy Ga:Ni molar ratio was found and employed for the characterization of the electroplated Ga-Ni layers. Results from the irradiations over natGa on Nb supports are presented in TABLE 2. TABLE 3 presents the results from irradiating two thick NiGa4 foils made by induction heating. Figure 1 contains the separation profile with DGA. 91% of the 68Ge is eluted in 2 mL of de-ionized water. We developed two simple methods for NiGa4 alloy manufacture. With a melting point > 800 °C and 80% presence of natGa, it is a more convenient target for 68Ge production compared to Ga encapsulated in Nb. The separation method based on the extraction resin DGA yields similar results as the liquid-liquid extraction method mentioned in [6,7], but we believe this is a more convenient method since it only requires a single trap-and-release step and not many extraction steps.

Introduction 52Mn (t½ =5.59 d, β+ = 29.6%, Eβmax = 0.58 MeV) has great potential as a long lived PET isotope for use in cell tracking studies, observation of immunologic response to disease states, or as an alternative to manganese-based MRI contrast agents. Its favorable max positron energy leads to superb imaging resolution, comparable to that of 18F.[1] Manganese is naturally taken up by cells via a multitude of pathways including the divalent metal transporter (DMT1), ZIP8, transferrin receptors (TfR), store-operated Ca2+ channels (SOC-Ca2+), and ionotropic glutamate receptor Ca2+ channels (GluR).[2] These natural transport mechanisms make 52Mn an attractive isotope for applications necessitating non-perturbative cell uptake. In particular, cell tracking is critical to the development and translation of stem cell therapies in regenerative medicine. Alternative-ly, 52Mn could be used in immunotherapy techniques such as adoptive cellular therapy (ACT) to evaluate the ability of external immune cells to reach their intended target. Material and Methods 52Mn was produced by natCr(p,x)52Mn using 16 MeV protons. The average thick target production yield was 0.23 mCi/µA-h with less than 0.25% co-production of 54Mn. Small amounts of 51Cr were observed in the target, but were absent from the radiochemically separated product. Target construction consisted of a water jet cooled chromium disc (3/4” diameter, 0.4” thick). Targets were purchased from Kamis Inc, and are 99.95% pure. Targets withstood beam currents of 30 µA with no visible aberration. Chromium targets were etched by concentrated HCl following bombardment. Mn2+ ions were extracted from 9M HCl to 0.8M trioctylamine in cyclohexane leaving the bulk chromium in the aqueous phase. After isolating the organic phase, 0.001M NH4OH was used to back-extract the Mn2+ ions to aqueous phase. This purification cycle was conducted a total of three times for each 52Mn production. Results and Conclusion For a starting bulk chromium mass of 456 ± 1 mg, a post-separation chromium mass of 5.35 ± 0.04 ng was measured by microwave plasma atomic emission spectrometry (MP-AES). This mass reduction corresponds to an average separation factor of 440 for a single purification cycle. Each purification cycle had a 52Mn recovery efficiency of 73 ± 7 % (n = 6), resulting in an overall separation efficiency of approximately 35 %. These efficiencies and separation factors agree reasonably well with the work conducted by Lahiri et. al.[3] Prior to use, the product was passed through a C-18 Sep-Pak to remove any residual organic phase. After four target irradiations and etchings, some pitting became noticeable on the target face. These have not yet compromised the o-ring seal with the target deplater, but it is possible that target replacement after every 6–9 52Mn productions will be necessary moving forward. Following the successful separation of 52Mn from chromium, in vitro experiments were conducted to demonstrate the uptake of 52Mn by human stem cells and mouse tumor cells. A linear uptake response was observed as a function of the amount of activity exposed to the cells for both cell models. These experiments have shown great promise for 52Mn as a long-lived PET isotope in cell tracking studies. Details will be presented.

Introduction The TR-24 is relatively new to the cyclotron market and its advantages over lower energy PET cyclotrons have not yet been fully realized. A new high current [18F] fluoride production target that takes advantage of the higher energy and current afforded by the TR-24 has been developed. Material and Methods The TR-24 cyclotron presents challenges of producing conventional PET isotopes even with its variable energy capability (18–25 MeV). Simultaneous irradiation of two targets that require different proton energies is possible only using beam energy degrader. Due to the relatively wide energy window, the degrader design is not trivial, especially for the high current operation. For example, reduction of beam energy from 24 to 18 MeV would require the use of an approximately 1.5 mm thick aluminum degrader. At 100 μA this degrader would have to be capable of dissipating 700 W of beam power, which would be challenging to achieve with no cooling or using a conventional helium cooling window. However, cooling water used as a beam energy degrader can dissipate several kilowatts of beam energy and provide additional cooling for target material and window foils. FIGURE 1 demonstrates the concept of the water cooled target window. A standard 18F- water target with a 2.5 mL fill volume and a 30 degree beam incident angle was modified to accept the new water window. A 1 mm thick region of circulating cooling water was inserted between the vacuum and the product foil. The combined beam energy degradation caused by the vacuum foil (0.00012“ Havar), the cooling water (1 mm) and the target foil (0.00012“ Havar) was approximately 7 MeV for a 24 MeV incident proton energy. The target was installed on a target selector mounted directly on the TR-24 cyclotron. No additional beam focusing or steering devices were used to defocus or correct beam shape. A small recirculation water system was setup to supply cooling water for the degrader. A mixed bed ion exchange column was installed on the return line to trap N-13 and radioactive metal ions that could possibly be etched from the Havar foils. The water in the degrader was continually circulated in a closed loop providing cooling to the vacuum and target foils. An 800mL/min water flow through the degrader was generated by a low pressure water pump. Results Several tests were performed with O-16 water to establish current – pressure curve and to determine “burn through” current (FIGURE 2). Conclusion Initial tests demonstrated that the new F-18 target with a 1 mm water degrader is capable of accepting power levels in excess of 3.6 kW, operating at 150 μA. More testing is under way, including testing with H218O to determine the F-18 production capacity of this target. We will look into adapting this concept to all ACSI PET targets, including the high current F-18 produc-tion target which can potentially reach an operational current of 200 μA.

Introduction The established methods for the production of 99Mo, based on fission in nuclear reactors, continue to present problems as a result of the plant’s aging and the significant investments needed for maintenance or for their renewal. Much research work is thus in progress on the study of alternative methods for the production of 99mTc in quantities and with the degree of purity required for the clinical use. Between them, the cyclotron production of 99mTc via the 100Mo(p,2n)99mTc reaction has turned out as the most attractive alternative. One critical aspect regarding the production of 99mTc with cyclotron is the need for a robust and reliable target production process. Several techniques have been indicated as extremely promising such has plasma spray and laser cladding; however these methods require specialized instrumentation and complex operations to be performed handling activated materials in order to recover irradiated Mo. In this work we report the development of the work done at the University of Bologna, as a part of a wider INFN project, as regards the methods of preparation of solid targets suitable for the production of 99mTc irradiating a target of 100Mo, employing a cyclotron for biomedical use, normally operated for the production of PET radionuclides. Material and Methods Irradiations were performed with a 16.5 MeV GE PETtrace cyclotron equipped with a solid target station previously developed by our group (1). In initial tests, a stack of 1–3 metallic foils, 100 μm thick, of natMo were irradiated with protons in the 15.9→9.8 MeV energy range. Foils were then dissolved in a HNO3-HCl solution and samples were analyzed with high resolution gamma-ray spectrometry (Canberra, including a HPGe detector with a 30% relative efficiency and a resolution of 1.8 keV at 1332 keV) using Genie2000 software; the measurement campaign lasted several weeks to take into account the different half-lives of the produced radionuclides. Results were extrapolated to a highly enriched 100Mo target and compared to Monte Carlo simulations previously performed with FLUKA and TALYS codes (2). In order to investigate a method of preparation of the target that would make easier the recovery of the enriched material and recycling for the preparation of a new target, it was subsequently studied the preparation of pellets of Molybdenum trioxide. MoO3 powder (Sigma Aldrich, 99.9% trace metals basis, particle size < 150 μm) was used to prepare pellets using a 10 ton press. Pellets obtained in this way were then sintered on a Platinum support using a CARBOLITE furnace under a controlled atmosphere; the temperature was ramped according to a controlled and reproducible temperature cycle. Sintered pellets were subjected to visual inspection, mechanical tests of resistance to loading and downloading in the cyclotron target station, thermal tests and then irradiated at increasing current. The irradiated targets were again visually inspected then weighed, dissolved and subjected to gamma-ray spectrometry analysis. Results and Conclusion The experimental saturation yield for 99mTc calculated on the basis of the gamma-ray analysis of irradiated metal foils, gave an extrapolated yield of 1.115 ± 0.015 GBq/μA for a 100 μm thick 100Mo enriched target, in accordance with the value of 1.107 ± 0.002 GBq/μA obtained in Monte Carlo simulations. On these bases, an irradiation of 3 h at 50 μA is expected to produce 16.3 ± 0.2 GBq of 99mTc; considering the use of an efficient purification system, a radionuclidic purity > 99.9 % 2 h after the EndOfBombardment and a specific activity comparable with the actual standards are expected as achievable. Experiments on sintering pellets are still on going at the time of writing this report; initial results showed that addition of proper aggregating materials allows for suitable pellets preparation. The sintering process allows to obtain pellets having sufficient mechanical strength to withstand loading and downloading operations. Initial irradiation tests with beam current up to 25 μA were performed successfully with no changes in mass and mechanical properties of the pellet. These encouraging results suggest that sintered pellets may be a relatively inexpensive and easy solution to prepare 100Mo targets for the cyclotron production of 99mTc. Further experimental tests at higher beam current will be performed in order to assess the maximum current achievable with no damage of the target. At the same time, a prototype automated module based on standard industrial components is in testing phase as regards performance in the separation and purification processes.

Introduction The established methods for the production of 99Mo, based on fission in nuclear reactors, continue to present problems as a result of the plant’s aging and the significant investments needed for maintenance or for their renewal. Much research work is thus in progress on the study of alternative methods for the production of 99mTc in quantities and with the degree of purity required for the clinical use. Between them, the cyclotron production of 99mTc via the 100Mo(p,2n)99mTc reaction has turned out as the most attractive alternative. One critical aspect regarding the production of 99mTc with cyclotron is the need for a robust and reliable target production process. Several techniques have been indicated as extremely promising such has plasma spray and laser cladding; however these methods require specialized instrumentation and complex operations to be performed handling activated materials in order to recover irradiated Mo. In this work we report the development of the work done at the University of Bologna, as a part of a wider INFN project, as regards the methods of preparation of solid targets suitable for the production of 99mTc irradiating a target of 100Mo, employing a cyclotron for biomedical use, normally operated for the production of PET radionuclides. Material and Methods Irradiations were performed with a 16.5 MeV GE PETtrace cyclotron equipped with a solid target station previously developed by our group (1). In initial tests, a stack of 1–3 metallic foils, 100 μm thick, of natMo were irradiated with protons in the 15.9→9.8 MeV energy range. Foils were then dissolved in a HNO3-HCl solution and samples were analyzed with high resolution gamma-ray spectrometry (Canberra, including a HPGe detector with a 30% relative efficiency and a resolution of 1.8 keV at 1332 keV) using Genie2000 software; the measurement campaign lasted several weeks to take into account the different half-lives of the produced radionuclides. Results were extrapolated to a highly enriched 100Mo target and compared to Monte Carlo simulations previously performed with FLUKA and TALYS codes (2). In order to investigate a method of preparation of the target that would make easier the recovery of the enriched material and recycling for the preparation of a new target, it was subsequently studied the preparation of pellets of Molybdenum trioxide. MoO3 powder (Sigma Aldrich, 99.9% trace metals basis, particle size < 150 μm) was used to prepare pellets using a 10 ton press. Pellets obtained in this way were then sintered on a Platinum support using a CARBOLITE furnace under a controlled atmosphere; the temperature was ramped according to a controlled and reproducible temperature cycle. Sintered pellets were subjected to visual inspection, mechanical tests of resistance to loading and downloading in the cyclotron target station, thermal tests and then irradiated at increasing current. The irradiated targets were again visually inspected then weighed, dissolved and subjected to gamma-ray spectrometry analysis. Results and Conclusion The experimental saturation yield for 99mTc calculated on the basis of the gamma-ray analysis of irradiated metal foils, gave an extrapolated yield of 1.115 ± 0.015 GBq/μA for a 100 μm thick 100Mo enriched target, in accordance with the value of 1.107 ± 0.002 GBq/μA obtained in Monte Carlo simulations. On these bases, an irradiation of 3 h at 50 μA is expected to produce 16.3 ± 0.2 GBq of 99mTc; considering the use of an efficient purification system, a radionuclidic purity > 99.9 % 2 h after the EndOfBombardment and a specific activity comparable with the actual standards are expected as achievable. Experiments on sintering pellets are still on going at the time of writing this report; initial results showed that addition of proper aggregating materials allows for suitable pellets preparation. The sintering process allows to obtain pellets having sufficient mechanical strength to withstand loading and downloading operations. Initial irradiation tests with beam current up to 25 μA were performed successfully with no changes in mass and mechanical properties of the pellet. These encouraging results suggest that sintered pellets may be a relatively inexpensive and easy solution to prepare 100Mo targets for the cyclotron production of 99mTc. Further experimental tests at higher beam current will be performed in order to assess the maximum current achievable with no damage of the target. At the same time, a prototype automated module based on standard industrial components is in testing phase as regards performance in the separation and purification processes.

No description available.